Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “REACTOR FUELING”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Evaluation of the energy balance of chemical looping combustion of solid fuels using CuO-based oxygen carriers

Chemical looping combustion (CLC) can inherently capture the CO2 generated during thermal energy production. When using CuO-based oxygen carriers, the reactions in both the fuel reactor (FR) and air reactor (AR) are exothermic, facilitating autothermal operation. To prevent the oxygen carrier from becoming agglomerated while ensuring a sufficient oxygen release rate, it is necessary to predict the FR temperature and remove an appropriate amount of heat from the AR when burning high-rank fuels. Here in this work, the energy balance of a CuO-based CLC system firing various solid fuels (biomass, coal, and petroleum coke) was investigated by establishing a thermodynamic model using FactSage. The energy distribution characteristics of the FR were quantified based on a virtual five-step thermodynamic sequence. The effects of fuel properties, oxygen carrier properties, and operating conditions on the energy balance were evaluated systematically. The results indicate that the temperature difference between the FR and AR is not very sensitive to the heating value of the fuel, while the CuO loading of the oxygen carrier, the heat capacity of the support material, and the flow rate of the fluidizing gas have significant impact. The insights obtained in this work will help improve the design and heat management of CLC.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Neutron Source Facility of the National Science Center “Kharkiv Institute of Physics and Technology” at Kharkiv, Ukraine

Argonne National Laboratory developed, designed, and supported the construction of a state-of-the-art Neutron Source Facility (NSF) at the Kharkov Institute of Physics and Technology (KIPT) in Kharkov Ukraine, under the U.S. DOE NNSA Russian Research Reactor Fuel Return (RRRFR) program. Ukraine approved the return of all highly enriched uranium in the country to Russia prior to the 2012 Nuclear Security Summit in return for the RRRFR program to fund the development, the design, and the construction of the NSF. The facility is designed to produce medical isotopes, train nuclear professionals, support the Ukrainian nuclear industry, and provide experimental capabilities for performing reactor physics, materials, and basic science research. The NSF was successfully started and operated in August 2021. The NSF is the first facility of this type in the world, and it will be used to understand the physics of driven systems for energy production and the disposal of spent nuclear fuels. The NSF consists of a subcritical assembly using low enriched uranium (LEU) fuel driven with an electron accelerator. The NSF target design utilizes tungsten or natural uranium for producing neutrons through photonuclear reactions using 100-MeV electrons. The accelerator power is 100 KW, which produces 3x10 14 neutrons per second from the uranium target. The subcritical assembly is designed to obtain the highest possible neutron flux intensity with an effective neutron multiplication factor of less than 0.98. Passive safety, reliability, and environmental considerations were included in the NSF design. The NSF utilization study shows that this neutron source has the capability of producing different medical isotopes. Several horizontal neutron channels are incorporated for performing applied and basic research. The NSF is configured to accommodate future design upgrades and new missions. The facility design was approved by the Ukraine Academy of Sciences before starting the NSF construction. The construction, the equipment installation, and the operation of the different systems of the NSF were completed and approved by the Ukrainian regulators. In July 2020, the State Nuclear Regulatory Inspectorate of Ukraine (SNRIU) issued a license for the physical startup of the NSF. The construction of facility was completed in early 2021.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Fuel Assembly and Irradiation Parametric Study for Extended-Enrichment and High-Burnup Light-Water Reactor Spent Nuclear Fuel in Dry Storage Casks and Transportation Packages

There is an increased interest in operating commercial light-water reactors (LWRs) in the United States with improved economics that would result from longer fuel cycle lengths, fewer and shorter refueling outages, and fewer fuel assemblies requiring storage at the back end of the fuel cycle. To support this, fuel discharge burnups, as well as initial 235 U enrichments, must be higher than those used in current commercial LWRs. The typical upper limit considered for assembly average burnup in this report is 75 gigawatt-days (GWd) per metric ton of uranium (MTU), as opposed to the current typical upper bound of approximately 62 GWd/MTU. The upper limit considered for initial 235 U enrichment is 8 weight percent (8 wt %), as opposed to the current regulatory limit of 5 wt %. The enrichment range from 5 to 8 wt % is referred to in this report as extended enrichment . To investigate the effect of high burnup and extended enrichment conditions on dose rates and burnup credit for dry storage casks and transportation packages, a fuel assembly and irradiation parametric study was performed. The conclusions from this study will assist U.S. Nuclear Regulatory Commission staff in reviewing applications for dry storage casks and transportation packages that contain high-burnup and extended enrichment fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Modeling, Performance Assessment, and Nodal Data Analysis of TRISO-Fueled Systems with Shift

This technical report documents several enhancements to the Shift Monte Carlo (MC) code under the US Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in fiscal year (FY) 2022. Performance enhancements were added to Shift specifically for tristructural isotropic (TRISO)–fueled reactor systems and guided based on performance analysis in FY 2021. For the pebble performance model developed in previous studies, the runtime improved by ~ 91× compared to the original model and ~ 2× compared to the user-optimized model. Compared to Serpent, Shift is ~ 3× slower if Serpent delta-tracking is enabled but ~ 2× faster when delta-tracking is disabled. The multigroup cross section generation was improved through simplifying tally input definitions, porting several post-processing tally operations from Python scripts into the Shift code base, and accounting for production reactions in the scattering multiplicity. Progress was also made on two emerging capabilities: (1) the development of Titan (a Shift reactor physics user interface) and (2) initial investigation into path-length tallies for computing multigroup scattering matrices.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Property Measurements of NaCl-UCl 3 and NaCl-KCl-UCl 3 Molten Salts (Rev.1)

Thermochemical and thermophysical property values of several salt compositions of interest are needed by molten salt reactor (MSR) developers to design, license, and operate their reactors. Thermochemical and thermophysical properties being measured at Argonne include thermal transitions, phase behavior, heat capacity, density, volumetric thermal expansion of the liquid phase, thermal diffusivity, thermal conductivity, and viscosity. Several properties of eutectic compositions in the ternary NaCl-KCl-UCl 3 and binary NaCl-UCl 3 systems that may be used by MSR developers as fuel bearing salts are being measured. A 65.8 mol % NaCl–34.2 mol % UCl 3 mixture and a near-eutectic mixture of 50.9 mol % NaCl–24.4 mol % KCl–24.7 mol % UCl 3 were synthesized and the thermochemical properties of the mixtures were measured by using differential scanning calorimetry (DSC). The measured transition temperatures were compared to transition temperatures predicted using two models. A thermodynamic model of the binary NaCl-UCl 3 system was constructed using data in the Molten Salt Thermal Properties Database–Thermochemical Version 2.0 (MSTDB-TC V2.0). A ternary NaClKCl-UCl 3 system model constructed at Argonne and was described in a previous report. These comparisons can be used to validate the models. Thermophysical property values of molten salts are needed to model how salt retains and transfers heat in an MSR system. These property values are essential to the entire MSR design because molten salt is used as both the fuel and the coolant material in a salt fueled reactor. Heat capacity of the synthesized NaCl-UCl 3 and NaCl-KCl-UCl 3 salt mixtures was measured by using DSC and thermal diffusivity was measured by using laser flash analysis (LFA) at temperatures spanning the typical operating range of an MSR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

ECAR-6076 Rev 0 MARVEL Reactor End of Life Enveloping Radiological Source Term

As part of the Microreactor Applications Research Validation and Evaluation Project (MARVEL) design effort, an enveloping end of life radiological source term for the reactor fuel and sodiumpotassium (NaK) eutectic coolant is calculated in support of the MARVEL preliminary hazard evaluation. The source term calculation assumes slightly more than two-years of uninterrupted operations at a power of about 120 kW for a total energy of 7,600,000 MJ which exceeds the planned two-year operation at 100 kW. The Reactor modeling was performed with MCNP6 Version 1.0 and the depletion was performed using the COUPLE and ORIGEN modules from the SCALE Version 6.2 code package.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2025

Infrastructure and resource development needs to support a significant expansion of US nuclear capacity

In December 2023, the United States pledged at a United Nations conference to triple nuclear energy by 2050. In May of 2025 a new administration announced plans to quadruple nuclear energy by 2025, setting a new target that requires understanding the key infrastructure and resources needed to achieve large-scale deployment. This study attempts to quantify “what it takes” to deploy an additional 200 GWe of nuclear capacity. It projects demand levels for workforce, fuel, uranium mining and enrichment, waste management, supply chain, land, licensing, and water. Demand growth is also contextualized. For example, the peak number of construction workers required is approximately 215,000, a 37 % increase in the current US utility construction workforce. Up to 80,000 MT/year of mined uranium will be needed, necessitating a 60-fold increase in US mining production or a doubling of global mining if sourced externally. Advanced reactor fuels, uranium mining, and nuclear-grade equipment production would need substantial expansion. Operational workforce, enrichment needs, and licensing submissions also face significant increases. Although water consumption may see relatively lower increases, securing water rights poses unique challenges. This study consolidates various demand metrics in a broad context, highlighting the necessity of early preparations, such as workforce training, to support a tripling of nuclear energy. While it outlines the demand-side requirements, it does not assess the difficulties of ramping up or other supply-side issues. The findings also shed light on the impact of reactor size and reactor technology on various demand metrics.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2022

TREAT Slide Decks for Alabama A&M

Slides to introduce TREAT and two senior design projects to a professor and senior engineering students at Alabama A&M. Slide Deck 1 is an overview of the TREAT Reactor and its mission. Use for Space Nuclear Thermal Propulsion Testing is mentioned, but no details of space reactor fuel are included. Review for unlimited release is requested. Slide deck 2 presents technical background for neutron radiography, with contrast to x-radiography, and then presents the application of TREAT for Flash Radiography. Slide Deck 3 describes the background behind the second senior design project, which is to design powered actuators for TREAT core clamps. Slide Deck 3 includes photos of different sides of the TREAT reactor and sketches of internal components associated with the core clamping system. These were presented via Teams on October 5, but we want to provide the slidedeck files to the university for student reference.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

MODELING A SPENT NUCLEAR FUEL CASK SEISMIC TEST

The US Department of Energy Spent Fuel and Waste Science and Technology (SFWST) program is planning to conduct a series of full-scale shake table tests to simulate hypothetical earthquake conditions and record the response of surrogate spent nuclear fuel (SNF) assemblies in a canister dry storage system mockup. The shake table motions will represent a range of hypothetical earthquake conditions at hypothetical locations in the continental US to generally define the range of mechanical loads that SNF can be expected to experience during extended dry storage periods. The earthquake conditions will represent seismic hazards in the 2,000-to-20,000-year return period range. The test will use instrumented pressurized water reactor fuel assemblies with surrogate mass inside the fuel rods instead of radioactive fuel pellets. Dummy assemblies with similar mass and dimensions to actual fuel assemblies will occupy the rest of the fuel assembly locations in a SNF canister that holds 32 fuel assemblies. The canister will be located inside a mockup vertical concrete cask. Instrumentation will record the motion of the major components of the complex dynamic system, and strain gauges will be used to record the cladding strain at select locations. Preparations for the test require modeling predictions to identify the range of response of the system and to help select specific earthquake cases to be simulated on the shake table from a large set of potential cases. This paper describes the pretest nonlinear finite element modeling efforts that have been completed to date, including cask system level modeling and fuel assembly modeling in LS-DYNA. The cask system level models are critical for anticipating sliding or tipping of an unanchored cask during the test. The fuel assembly model is needed to estimate the range of cladding strain response and fuel assembly structural response to be expected during the test and cladding strain measurements will be key metrics for model validation and the development of modeling best practices after the test is concluded.

Klymyshyn, Nicholas A.↗
DOE OSTI2023

Status of Mechanistic Fission Gas Model in High-Burnup Fuel

A desire to increase fuel burnup to decrease the cost of nuclear power plants has led to significant interest within the nuclear industry to develop improved understanding of high-burnup nuclear fuel microstructure and the potential for fuel fragmentation, relocation, and dispersal that contribute to burnup and safe operating limits. This milestone report describes joint research activities and program planning to develop mechanistic models for high-burnup UO 2 microstructure, including both intra- and intergranular gas bubble populations and fission gas release, specifically associated with transient release. This model development is being extensively leveraged against a rapidly growing experimental database of high-fidelity electron microscopy characterization of commercial, light water reactor fuel in the as-irradiated condition as well as that following simulated loss-of-coolant test conditions. This report describes the status of model development, highlights recent microstructural data, and summarizes the data needs to complete initial development and experimental validation of mechanistic models of fission gas and microstructural evolution at high burnup, as well as transient fission gas release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Development of Shielding Benchmarks Using the Godiva IV Assembly

The Nuclear Criticality Safety Program (NCSP) is developing a shielding benchmark using the Godiva IV assembly as a source. Its present status is reviewed herein. Even eight decades into the nuclear era, substantial work remains to develop a database of shielding benchmarks to support future nuclear development. A nuclear simulation is only as good as its supporting data, inputs, and validation basis. Uncertainty in these areas is addressed using conservatism, which adds margin and, occasionally, cost. In many shielding situations, high accuracy is not necessary because additional material is not particularly expensive. After all, 1–2 cm of lead often reduces the gamma ray dose substantially. However, in certain areas, conservatism can add unnecessary cost. These areas include mobile shielding applications such as casks, ships, microreactors, and spacecraft, where weight and, thus, margin is expensive. Although these characteristics are side benefits for NCSP shielding benchmark development, the main driver is enabling more reliable placement of criticality accident alarm systems (CAASs) in nuclear material facilities, such as those dedicated to the production of advanced reactor fuels. CAAS placement relies on more than accurate data and code validation. It also relies on sufficiently accurate materials specifications, geometry specifications, and a well-defined, alarm-producing baseline accident. All these things require tacit knowledge and understanding of the problem being evaluated. Benchmarks can help ensure this understanding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

HALEU Experiments [Slides]

The DNCSH, through the HALEU Availability Program, needs publicly available benchmarks supporting Advanced Reactor Fuel Cycle and Transportation. LANL is uniquely positioned to support this goal, with its past experiments at TA-18, planned experiments at NCERC, experiment design capability, and ties to industry. The Deimos experiment will be executed soon and serve as a premier benchmark for the program. LANL expects to submit proposals to the call when issued (soon). Additional experiments supporting DNCSH have already been designed

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Microstructure of irradiated AGR TRISO particle buffer layers as measured by X-ray computed tomography

Shrinkage of the initially low-density buffer layer in tristructural isotropic (TRISO) coated fuel particles during irradiation is a well-known phenomenon with potential implications for fission product and actinide transport, as well as potential pyrocarbon fracture that in some cases has been observed to impact particle performance. During post-irradiation examination, the buffer layer's structure is commonly determined using 2D microscopy of a particle cross section or a series of particle cross sections at staggered depths. Although these methods provide a general idea of the irradiated buffer microstructure, they do not provide a full picture of the TRISO coating microstructure. By contrast, x-ray computed tomography (XCT) provides full 3D imaging of the TRISO particle. Particles from various compacts irradiated for the Advanced Gas Reactor Fuel Development and Qualification Program were imaged using XCT. Some of these particles were selected because of abnormal fission product inventories (e.g., low 137 Cs), whereas others were randomly selected from the center of the fission product inventory distribution. Qualitative and quantitative analysis techniques were applied to the randomly selected particles as representatives of typical TRISO behavior to study the post-irradiation structure of the buffer layer. These results showed that, while separation of the buffer and inner pyrolytic carbon layers was a common behavior during the radiation-induced shrinkage of the buffer, a portion of the original buffer/inner pyrolytic carbon interface remained intact throughout irradiation in nearly all cases. Furthermore, these results also indicated clear trends in the degree of buffer densification with irradiation temperature and fluence.

36 MATERIALS SCIENCE↗
DOE OSTI2020

ECAR-5139 MARVEL Radiological Source Term

As part of the Microreactor Applications Research Validation and Evaluation Project (MARVEL) design effort, a preliminary end of life reactor fuel radiological source term is calculated in support of the MARVEL preliminary hazard evaluation. The source term calculation assumes a two-year uninterrupted operating life at a power of about 111 kW for a total energy of 7,000,000 MJ. A final source term for the final hazard evaluation will need to be generated or an evaluation performed to determine the applicability of the source term documented here for use in the final hazard evaluation.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Development of pebble-based extruded carbon rods for extreme plasma heat flux environments

This work presents first experiments toward the development of continuously renewable (extrudable) pebble-based carbon rods for use as plasma-facing components in extreme steady-state plasma flux environments. The primary envisioned application of this work is a first wall that can survive long-term in future magnetic fusion power reactors while also improving recovery of the reactor fuel (tritium and deuterium atoms). Bench tests applying extreme steady-state front-surface heat loads of up to 50 MW/m 2 are presented. Continuous pebble rod front-surface recession and intact pebble recovery are successfully demonstrated, at a rate of order 0.2 cm/s. Numerical simulations of the pebble rod front-surface recession are able to match observations reasonably well, indicating that the recession mechanism can be understood as occurring due to pebble thermal expansion and resulting shock and cracking of the inter-pebble binder. Tests of the pebble rod extrusion demonstrate that friction between the rods and the stainless-steel extrusion channel is tolerably low (<50 N for the expected channel length) over a wide range of temperatures. Front-surface outgassing rates below 1000 Torr L/s/m 2 are achieved, believed to be sufficiently low for use in magnetic fusion reactors. In conclusion, initial parametric scans over pebble rod size and binder fraction to vary front-surface recession rates are presented.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Linear source method of characteristics in HELIOS2

The Studsvik Scandpower code HELIOS2 is a fast lattice transport code that has gained wide use for reactor fuel cycle analysis in commercial, regulatory, and research environments. It contains two transport solvers - Collision Probabilities (CP) and Method of Characteristics (MoC) - with a general geometry ability and generalized depletion for a variety of fuel types, experimental facilities, and isotope production applications. The HELIOS2 lattice code has been extended with the development of a linear source MoC option. The linear source equations are derived within the scope of the HELIOS2 general geometry and non-cyclic ray tracing. The implementation was tested for fine and coarse spatial mesh models of the 2D C5G7 MOX computational benchmark, which provides a pure and extreme test of a transport solver. The linear source yields at least comparable accuracy with reduced computational resource requirements relative to the flat-source approximation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Momentum informed muon scattering tomography for monitoring spent nuclear fuels in dry storage cask

Development of an effective monitoring method for spent nuclear fuel (SNF) in a dry storage cask (DSC) is important to meet the increasing demand for dry storage investigations. The DSC investigation should provide information about the quantity of stored SNF, and quality assurance of materials should be possible without opening the cask. However, traditional nondestructive examination (NDE) methods such as x-rays are difficult to deploy for DSC investigation because a typical DSC is intentionally designed to shield against radiation. To address this challenge, cosmic ray muons (CRMs) are used as an alternative NDE radiation probe because they can easily penetrate an entire DSC system; however, a wide application of muons is often hindered due to the naturally low CRM flux (~10 4 muons/m 2 /min). This paper introduces a newly proposed imaging algorithm, momentum-informed muon scattering tomography (MMST) , and presents how a limitation of the current muon scattering tomography technique has been addressed by measuring muon momentum. To demonstrate its functionality, a commercial DSC with 24 pressurized light water reactor fuel assemblies (FAs) and the MMST system were designed in GEANT4. Three noticeable improvements were observed for MMST system as a DSC investigation tool: (1) a signal stabilization, (2) an enhanced capability to differentiate various materials, and (3) statistically increased precision to identify and locate missing FAs. The results show that MMST improves the investigation accuracy from 79 to 98% when one FA is missing and 51% to 88% when one-half FA is missing. The advancement of the NDE technique using CRM for DSC verification is expected to resolve long-standing problems in increasing demand for DSC inspections and nuclear security.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Stress rupture of SiC/SiC composite tubes under high-temperature steam

SiC-fiber–reinforced SiC matrix composite cladding for light water reactor fuel elements must withstand high-temperature steam oxidation in a loss-of-coolant accident scenario (LOCA). Current composite designs include an outer monolithic SiC layer, in part, to increase steam oxidation resistance. However, it is not clear how such a structure would behave under high-temperature steam in the case when the monolithic layer cracks and carbon interphases and SiC fibers are exposed to the environment. To fill this knowledge gap, stress-rupture tests of prototypic SiC composite cladding at 1000°C under steam and inert environments were conducted. The applied stress was ~120 MPa, which was beyond the initial cracking stress. The failure lifetime under steam was 400–1300 s, while 75% of the composite specimens did not fail after 3 h of total exposure under inert gases. Microstructural observations suggest that steam oxidation activated slow crack growth in the fibers, which led to failure of the composite. The results from this study suggest that stress rupture in steam environments could be a limiting factor of the cladding under reactor LOCA conditions.

36 MATERIALS SCIENCE↗